Concentration of pu using oxalate type carrier



2,933,369 Patented Apr. 19, 1960 CONCENTRATION OF Pu USING OXALATE TYPE CARRIER David M. Ritter and Robert P. S.-Black, Oak Ridge, Tenn., assignors to the United States of America as represented by the United States Atomic Energy Commission Application September 28, 1944, Serial No. 556,156 Claims. (Cl. 213-145) This invention relates in general to methods for processing materials containing the element of atomic number 94, known as plutonium, and specifically to a method for separating the plutonium from extraneous matter such as substances of the kind present in neutron irradiated uranium as exemplified by uranium and especially fission products, and the like radioactive contaminants. More particularly, this invention concerns a separatory and concentration procedure involving the use of an oxalate carrier wherein certain improved steps may be employed in treating the carrier.

As described herein, the isotope of element 94 having a mass of 239 is referred to as 94 and is also called plutonium, symbol Pu. In addition, the isotope of element 93 having a mass of 239 is referred to as 93 Reference herein to any of the elements or the term plutonium value is to be understood as denoting the element generically, whether in its free state or in the form of a compound, unless indicated otherwise by the context.

Elements 93 and 94 may be obtained from uranium by various processes which do not form a part of the present invention including irradiation of uranium with neutrons. Neutron irradiated uranium may be prepared by reacting uranium with neutrons from any suitable neutron source, but preferably the neutrons used are obtained from a chain reaction of neutrons with uranium.

Naturally occurring uranium contains a major portion of U a minor portion of U and small amounts of other substances such as UX and UX When a mass of such uranium is subjected to neutron irradiation, particularly with neutrons of resonance or thermal energies, 'U by capture of a neutron becomes U which has a half life of about 23 minutes and by beta decay becomes 93 The 93 has a half life of about 2.3 days and by beta decay becomes 94". Thus, neutron irradiated uranium contains both 93 and 94 but by storing such irradiated uranium for a suitable period of time, the 93 is converted almost entirely to 94 In addition to the above-mentioned reaction, the reaction of neutrons with fissionable nuclei such as the nucleus of U results in the production of a large number of radioactive fission products. As it is undesirable to produce a large concentration of these fission products which must, in view of their high radioactivity, be separated from the 94 and further as the weight of radioactive fission products present in neutron irradiated uranium is proportional to the amounts of 93 and 94 formed therein, it is preferable to discontinue the irradiation of the uranium by neutrons when the combined amount of 93 and 94 is equal to approximately 0.02 percent by weight of the uranium mass. At this concentration of these substances, the concentration of fission elements which must be removed is approximately the same percentage.

A number of processes have already been proposed for accomplishing the separation and concentration of Pu. Certain of these processes are generically known as the bismuth phosphate process and the wet fluoride" process. These processes are the invention of others and the details of the processes are described in copending applications, as for example application Serial No. 519,714, new Patent No. 2,785,951, to be referred to hereinafter, which gives complete details relative to the bismuth phosphate process. Still another type of procedure is the oxalate method wherein the precipitate which carries Pu is an oxalate compound. Consequently,

all of the details of operation of the aforementioned processes are not described herein.

While the aforementioned processes have been used successfully, the present invention provides certain variations in these processes for obtaining additional advantages. That is, by the present invention utilizing many of the same steps which have been previously worked out, but supplemented and altered in certain respects the advantage of easier reaction may be obtained as will be' described.

The term decontamination as used herein refers in particular to the separation or removal of radioactive material from the Pu, or materials containing thePu. The meaning of the term carrier precipitate and other similar terms will be apparent as the description proceeds.

This invention has for one object, to provide a new method for the separation and recovery of plutonium.

Another object is to provide a method of separating plutonium by a procedure wherein in certain steps different reagents than have heretofore been used are employed.

A still further object is to provide a process for separat ing Pu wherein elimination of extraneous materials such as fission products may be better accomplished.

Still another object is to provide a novel process for separating the Pu which lends itself to the use of and to coupling with steps already known or practiced.

Another object is to provide a separatory process for the recovery of Pu which may employ many of the materials used in existing processes and which may be carried out in existing equipment without change, or with a minimum of equipment change.

Still another object is to provide a process for the recovery of Pu which may be applied to solutions containing Pu either in relatively small or large amounts.

' Another object is to provide a separatory and concentration process for Pu containing materials involving the use of both oxalate and bismuth phosphate carriers.

Still another object is to provide a method of obtaining an acid solution of materials that resist direct dissolution in acid.

Other objects will appear hereinafter.

For a further understanding of the invention, reference will later be made to the attached drawing forming a part of the present application. In this drawing, a diagrammatic representation of one embodiment of the invention is given in the form of a flow sheet.

We have found that Pu in admixture with various extraneous material may be separated and concentrated by the use of the series of steps involving the utilization of a bismuth phosphate treatment followed by an oxalate treatment. These treatments are in accordance with known practice. However, we have found that the combination of steps may be rendered simpler and more efficient by interposing the improvement features described herein such as a different type of dissolution procedure or optionally the use of different type oxidizing agents and other variations as set forth in detail.

By the utilization of these features, not only are the advantages previously obtained in the processes still obtainable, but easier and simplified operation as well as savings in time are infected.-

The bismuth phosphate process is set forth in applicaout of solution.

tionSferial Number 51 9,7151, now Patent No. 2,785,951, filed January 26, 1944, Thompson and Seaborg, and ref: erence is made to that application for full disclosure of such process, details thereof being omitted from the present disclosure. except Where necessary to an understanding of the present invention. As set forth in said application, it has been discovered that plutonium has more than one oxidation state, including a lower oxidation state or states principally the tetravalent state referred to herein as Pu in which the element is characterized by forming insoluble phosphates and fluorides and a. higher oxidation state or states principally the hexavalent statereferred to as Pu in which theelement forms soluble phosphates and fluorides. (In general, similarremark ppl to p ss so e tin vxshtepmasses; that is, Pu is characterized byforminginsolub-leoxalate,- while;Pu forms solublegoxalatet), Thus; the abi ity f -.pluto um in e-pre ce of phosphateon is llangedby changing the oxidationstate ofthe, plu-,

toni um. As foreign product, such as-uraniumand ura nium fission; products, which are normally encountered as contaminants in plutonium production, form substanceswhich are in part insoluble and in part soluble inthe presence of phosphate ions, the process contemplates alternately separating, by precipitation, plutonium in its phosphate insoluble lower oxidation state from the phosphate soluble foreign products, and with the plutonium in its higher oxidation (phosphate soluble) state, separating the plutonium from foreign products which are in a phosphate insoluble state. However, where the 30 plutonium is present in relatively low concentrations, as,

for example, 10 milligrams per liter, as is often encountered, it is below its limit of solubility, as phosphate, in

the solution, so that the addition of phosphate ions will normally not'cause the bulkof the plutonium to be thrown down as a precipitate. In such case it has been found desirable to use suitable substances which will be converted intoinsoluble phosphate carriers by phosphate ions and will thereupon carry the plutonium phosphate In this regard, bismuth phosphate, in

particular, has been found to be a very effective carrier for this purpose. Bi+ ions may be introduced into the, solution (e.g. a l N HNO solution) by the addition of bismuth nitrate or other bismuth compound, preferably as an acid solution; usually, 10 to 25 mg. of Bi+ per 10 cc. of solution is sufiicient to precipitate with phosphate ions and carry substantially all of the plutoniumv out of solution. An excess-of phosphoric acid is added to thesolution to precipitate the-bismuth and plutonium as insoluble phosphates; a phosphoric acid concentration in the solution of approximately 0.4 M to 0.8 Mis usually satisfactory. Alternatively, the Bi+ may beintroduced intothe solution already containing the the phosphoric acid. Preferably, the solution may be heated for approximately an hour at about 75 C. to promote substantially complete precipitation. To insure the presence of plutoniumin said reduced state during thecarrier precipitation, it may be desirable to initially add small amounts of a reducing agent, e.g. ferrous salts, to the solution. The resulting bismuth phosphate precipitate carrying plutonium values is then separated from the solution by such means as filtration, centrifugation, or

the like, and the separated precipitate many then be dissolved in an excess of a concentrated acid, e.g.- 10 N: nitric acid.

The aforementioned processes are conducted under acid conditions. It has heretofore been customary to dissolve the-oxalate carrier precipitate in an acid, such as nitric acid, to obtain the aforesaid acid conditions. How.-

ever, it has been observed that the oxalate precipitate dis- 70 solves with difficulty and that dissolution consumes a relatively long period of time.

We have discovered, as will be described in detail hereiuafter, that, possibly this ,difliculty may be dueto carryover of bismuth ions or some similarmaterials. fromnthe preceding steps in the process: We have found that this difi'iculty of redissolution of oxalate precipitate carrying Pu may be readily overcome by utilizing certain alkali reagents for dissolving the oxalate precipitate. Not only is the dissolution of oxalate precipitate materially facilitated by the use of such solvents, but as will be further described in detail, under the alkaline conditions of such dissolution, it is possible to utilize as oxidizing agents materials not hitherto regarded as capable of such use. The resultant alkaline solutions of the oxalate precipitates may then be readily rendered acidic and subjected to further treatment.

In general, the operation of our process for the separation and recovery of Pu from extraneous materials is as follows: Materials, which have undergone neutron bombardment and otherwise processed to contain Pu are given conventional preliminary treatments and finally a BiPO plutonium product precipitation is carried out.

The--,BiP 04,. prccipitatecarryingPuis dissolved in acid, the solutiongtreatedwith oxalic acid and, optionally, With-KC1, Uranous oxalate is precipitated carrying Pu by adding a source of U in H 80 solution. The uranous oxalate precipitate is dissolved in a soluble carbonate solution or in alkali, oxalate solution, in accordancewith the present invention, and the solution is oxidized. Hydroxylamine may be used to eifect oxidation in the. alkaline solution. The oxidized solution is neutralized with acid. Thereafter HNO HCl and H 80 may be added, and a second uranous oxalate product precipitation effected. This second precipitate may be dissolved in HNO and further treated as desired as by lanthanum fluoride precipitation to concentrate the product.

A sti l furtherunderstanding of the general application 5 of our, process may be had by reference to the attached flowsheet, The sourcefrom which the plutonium is to be separatedand recovered is indicatedat I. This source subiectedto standard extraction and decontamination procedure such as the bismuthphosphate procedure as'in- 0 dicated atII. The-precipitate of bismuth phosphate carrying' Pu obtained from, the foregoing procedure may be dissolved in acid in the conventional manner at III. Likewise, the formation of an oxalate precipitate'for carrying Pu may be carried out at IV in accordance with conventionalprocedure. However, the oxalate precipitate carrying- Pu resulting from precipitation at IV presents ditficulty of dissolution.

In accordance with the present invention, at V, special dissolution procedure isapplied. If desired, special oxidation procedure in accordance with the present invention may be applied at VI.

The foregoing operations produce a' solution to whichmay now be applied a standard oxalate precipitation operation at VII. The oxalate precipitate carrying Pu from VII'may be redissolved at VIII by conventional solvents such as nitric acid. Thereafter at IX standard operations for, further separating and concentrating Pu may be applied.

As apparent from the foregoing by the interposing of the steps of the present invention, the use of an oxalate type of process with a phosphate type of process is facilitated. Wehave provided a-procedureby which an acid,

. to a more detailed example, the steps making up the several'operations are set forth below.

Example The materials containing Pu which had undergone neutron treatment were dissolved and conventional preliminary treatment applied. The solution from the preceding step was reduced by any of the several reducing procedures. For example, hydrogen peroxide, ferrous salts and the like may be used as described in application Ser. No. 519,714, now Patent No. 2,785,951, aforementioned. The standard BiPO precipitation procedure was carried out on the solution in (r) condition to obtain a Pu containing product precipitate. The foregoing steps are embraced under operations I to III inclusive of the I Referring now to the steps pertaining in particular to this invention, the bismuth phosphate precipitate containing Pu from the preceding treatments was further processed as follows: About 1350 ml. of solution were prepared by dissolving the BiPO4 precipitate in 7 N HCl. The solution contained approximately 150 g'rs. Bi+++ per liter. The solution was then treated to contain .4 M of oxalic acid and about .1 M of KCl. The addition of the latter is optional. Uranous oxalate was precipitated by adding approximately a gram of U++++ as a .05 M solution in .1 N H 80 This addition was made over a period of 6 hr. at room temperature. The mixture was digested for about 1 hr. An oxalate precipitate carrying Pu was separated.

If it is attempted to dissolve the aforementioned oxalate precipitate in HNO it requires several hours heating at 75 C. before the precipitate starts to dissolve. On the other hand utilizing our invention employing alkaline reagents exemplified by alkali metal carbonates and alkali metal oxalates, the precipitate dissolves in a few minutes, as indicated in the next step.

The aforementioned uranous oxalate precipitate carrying Pu was readily dissolved in 4.2 ml. of 50% K CO solution. The carrier in the resultant solution was oxidized by adding .66 ml. of H 0 solution. We have also found that hydroxylamine may be used as an oxidizing agent in lieu of the peroxide. We have found hydroxylamine has the property of acting as an oxidizing agent in the aforementioned alkaline environment, but does not oxidize in acid solution. It will be observed here that the function of the oxidation is to convert the uranous oxalate to uranyl oxalate. In other words, the oxidizing agent merely functions to change the oxidation state of the carrier. The Pu is not oxidized.

After oxidation had been accomplished, there was added 2.9 ml. of 16 N HNO to neutralize the K CO Then the following additions were made: 3.7 ml. further of HNO 4.5 ml. (12 N) HCl; .84 ml. (36 N) H 80 The resultant solution was made up to 30 cc. volume by adding Water. A second uranous oxalate precipitate, carrying Pu was obtained in the aforementioned solution by making an addition of U++++ over a period of approximately /1 hr. in the amount of 2 grams per liter in a manner comparable to the procedure set forth above for forming the first oxalate precipitate. This second precipitate was separated and dissolved in HNO and further treated in a conventional manner by lanthanum fluoride precipitation, to further concentrate the product.

While the process has been described in particular as being carried out utilizing the uranous oxalate type of carrier, our invention is not limited specifically to this carrier. There are various other oxalate types of carriers known to carry Pu such as thorium oxalate and lanthanum oxalate. Our invention may be usedin a similar manner for facilitating operation in other instances where the carrier resists acid dissolution.

As described above, it is possible that the resistance to dissolution of such oxalate carriers may be caused by certain metallic ions carried along from preceding steps in the process. However, by using the alkaline solvents described herein, it has been found that dissolution may be rapidly accomplished and also that the desired acid conditions for further precipitation may be obtained.

It will be further observed that although the alkali dissolution conditions are employed for placing the first of the oxalate carrier precipitates in solution that in subsequent dissolution steps applied to oxalate precipitates, the dissolution may be directly accomplished with conventional solvents such as nitric acid as may be noted in the preceding example. However, if desired, all oxalate carrier dissolution steps may be carried out using our alkaline solvents. r.

Any of the soluble forms of the alkali reagents referred to herein such as sodium, potassium and ammonium salts may be used. The particular amount of or strength'of reagent used is not a limitation on our invention. In general, in processes of the type described, since reduction of volumes and concentration is desired, the use of the smallest amount of alkaline reagent consistent with good dissolution would preferably be employed. How ever, this does not preclude the use of larger amounts of reagents or diluted reagents as the process may be operated with excesses.

While the solution treated in the example described herein contained a small amount of Pu this is not a limitation as our process may be carried out on solutions containing various amounts of Pu from, for example, tracer amounts to 260 grams per liter. The precipitates may be separated by centrifuging or filtering and other conventional techniques may be employed in carrying out the process.

It is to be understood that all matter contained in the above description and examples shall be interpreted as illustrative and not limitative of the scope of this invention, and it is intended to claim the present invention as broadly as possible in view of the prior art.

We claim:

I. In a process for the recovery of plutonium values from an aqueous solution containing the same comprising precipitating uranous oxalate therein to thereby carrier precipitate said plutonium values from solution therewith, the improved method for subsequent processing which comprises contacting the resulting oxalate carrier precipitate with an aqueous solution of a reagent chosen from the group consisting of the carbonates and oxalates of alkali metals and ammonium to thereby dissolve the oxalate precipitate together with said plutonium values carried therein, thereupon incorporating in the resulting solution an oxidant selected from the group consisting of hydroxylamine and hydrogen peroxide to thereby oxidize dissolved uranous ions to the uranyl oxidation state, thereafter acidifying the solution, and providing a source of uranous ions therein thereby precipitating uranous oxalate to thereby carrier precipitate said plutonium values from solution therewith.

2. The process of claim 1 wherein said alkali metal carbonate is specifically potassium carbonate.

3. The process of claim 1 wherein said oxidant is specifically hydroxylamine.

4. The process of claim 1 wherein said oxidant is specifically hydrogen peroxide.

5. In a process for the recovery of plutonium values from an aqueous solution containing the same comprising precipitating bismuth phosphate therein to thereby carrier precipitate said plutonium values from solution therewith, then dissolving the resulting carrier precipitate to obtain an aqueous solution of said plutonium values, and thereafter precipitating in the obtained solution uranous oxalate to thereby carrier precipitate said plutonium values from solution therewith, the improved method for subsequent processing which comprises contacting the resulting uranous oxalate carrier precipitate with an aqueous solution of potassium carbonate to thereby dissolve the oxalate precipitate together with said plutonium values carried therein, thereupon incorp'orating hydrogen peroxide in the resulting alkaline solution to oxidize: dissolved uranous ions to the uranyl oxidation state, thereafter acidifying the resulting solution with an acid selected from the group consisting of nitric, hydrochloric, and sulfuric acids, and providing a source of uranous ions therein thereby precipitatinguranous oxalate to thereby carrier precipitate said plutonium values from said solution therewith.

References Cited in the file of this patent UNITED STATES PATENTS 2,206,634 Fermi et a1. July 2, 1940 i E OTHER REFERENCES l iohlse hutter:' Berichteder Deutsche Chemisclrie Gesellschaft, vol. 34, page 3623 (1901).

Marchi et al.: Journal of the'American Chemical Society, vol. 5, page 333 (March 1943).

Chi-1702, US. Atomic Energy Commission Report for month ending June 1, 1944; pages 7-12. Available only from Atomic Energy Commission on microcard. 

1. IN A PROCESS FOR THE RECOVERY OF PLUTONIUM VALUES FROM AN AQUEOUS SOLUTION CONTAINING THE SAME COMPRISING PRECIPITATING URANOUS OXALATE THEREIN TO THEREBY CARRIER PRECIPITATE SAID PLUTONIUM VALUES FROM SOLUTION THEREWITH, THE IMPROVED METHOD FOR SUBSEQUENT PROCESS WHICH COMPRISES CONTACTING THE RESULTING OXALATE CARRIER PRECIPITATE WITH AN AQUEOUS SOLUTION OF A REAGENT CHOSE FROM THE GROUP CONSISTING OF THE CARBONITES AND OXALATES OF ALKALI METALS AND AMMONIUM TO THEREBY DISSOLVE THE OXALATE PRECIPITATE TOGETHER WITH SAID PLUTONIUM VALUES CARRIED THEREIN, THEREUPON INCORPORATING IN THE RESULTING SOLUTION OF AN EXODANT SELECTED FROM THE GROUP CONSISTING OF HYDROXYLAMINE AND HYDROGEN PEROXIDE TO THEREBY OXIDIZE DISSOLVE URANEOUS IONS TO THE URANYL OXIDATION STATE, THEREAFTER ACIDIFYING THE SOLUTION, AND PROVIDING A SOURCE OF URANOUS IONS THEREIN THEREBY PRECIPATING URANOUS OXALATE TO THEREBY CARRIER PRECIPITATE SAID PLUTONIUM VALUES FROM SOLUTION THEREWITH. 